1. Industrial Field of the Invention
The present invention relates to a fuel assembly which constitutes a core of a nuclear reactor and, more particularly, to a fuel assembly for a light water reactor whose fuel is plutonium-uranium mixed oxide, and a core of a light water reactor utilizing the same.
2. Description of the Related Art
One conventional example of a fuel assembly used for a core of a nuclear reactor is an 8.times.8 type fuel assembly for a boiling water reactor (BWR). This fuel assembly is composed of a bundle of rods comprising a large number of elongated cylindrical fuel rods and water rods. The fuel rod is constituted of a fuel cladding which is filled with a large number of cylindrical UO2 fuel pellets and sealed both at upper and lower ends. The water rod has a cooling water inlet formed in a lower portion thereof and a cooling water outlet formed in an upper portion thereof so that cooling water flows inside of the water rod from the lower portion to the upper portion.
FIG. 21 is a horizontal cross-sectional view showing one example of an 8.times.8 type uranium fuel assembly according to the conventional technique.
As shown in FIG. 21, the fuel assembly 190 comprises fuel rods 192 of uranium oxide which are arranged in 8.times.8 lattice form, two water rods 196 which are located in a central portion thereof, and a channel box 197 which surrounds the fuel rods 192 and the water rods 196. When the fuel assembly 190 is mounted in a reactor core, a position adjacent to one corner portion of the fuel assembly 190 is used as a space in which a cruciform control rod 191 is inserted.
FIG. 22 shows another example of an 8.times.8 type uranium fuel assembly according to the conventional technique. In this fuel assembly 200, one water rod 206 is provided in the center of an 8.times.8 arrangement of fuel rods. The water rod 206, which is a large water rod, is located in an area from which four fuel rods 192 are removed. Except for this water rod, the fuel assembly 200 has substantially the same structure as the above-described fuel assembly 190.
In these two conventional examples, it is necessary to suppress the power peaking of the fuel assembly to a predetermined value or less and to prevent power of the fuel assembly per unit length, i.e., a linear heat generation ratio, from exceeding a limit value, in order not to deteriorate the fuel integrity.
On the other hand, from the standpoint of effective utilization of uranium resources, there has recently been carried forward a plutonium utilization in LWR program in which plutonium in used uranium fuel taken out of a light water reactor is recycled to the light water reactor. In this program, part or most of uranium fuel rods in a uranium fuel assembly are substituted by fuel rods of mixed oxide (MOX) in which plutonium is enriched, and MOX fuel assemblies thus obtained are mounted, as replacement fuel, on a light water reactor with uranium fuel assemblies. In this case, properties of the MOX fuel assembly should preferably be similar to those of the uranium fuel assembly. Also, the design of uranium fuel has a high-burnup tendency, and accordingly, the design of MOX fuel should preferably have a high-enrichment tendency, i.e., a plutonium load per fuel assembly should preferably be made as large as possible. However, when a load ratio of plutonium of the MOX fuel assembly is increased, there is caused a difference between reactor core properties of the MOX fuel assembly and those of a uranium fuel assembly owing to a difference between nuclear properties of uranium and those of plutonium. That is to say, because plutonium 239 Pu, 241 Pu which is fissile material has a thermal neutron absorption cross section larger than uranium 235 U, and because a neutron resonance effect by plutonium 240 Pu is larger than uranium 235 U, a neutron flux spectrum of MOX fuel is harder than a neutron flux spectrum of uranium fuel, thereby deteriorating neutron moderation. As a result, there are caused a decrease in thermal margin in a transient event due to an increase in an absolute value of void reactivity coefficient, an increase in distortion of the axial power distribution, a decrease in a moderator reactivity coefficient, or a decrease in reactor shutdown margin due to a decrease in a control rod worth.
Deteriorations of these properties are in allowable ranges if a fissile plutonium load is about 1/3 or less of the whole fissile material, and consequently, the structural design of the uranium fuel assembly can be used as it is. However, if the fissile plutonium load is higher, a water to fuel ratio of the fuel lattice must be increased to improve the neutron moderation.
The following two examples are known as means for improving the neutron moderation of the MOX fuel assembly:
(1) Fuel Assembly for Nuclear Reactor (Japanese Patent Unexamined Publication No. 63-172990)
In this conventional technique, four fuel rods in the vicinity of one large water rod provided in the center of an MOX fuel assembly are substituted by four water rods having the same diameter as the fuel rods, so that the absolute value of the void reactivity coefficient of the MOX fuel assembly will be decreased.
(2) Boiling Water Reactor (Japanese Patent Unexamined Publication No. 63-293493)
In this conventional technique, the water to fuel ratio of an MOX fuel assembly is made larger than that of a uranium fuel assembly by increasing the diameter of a water rod located in the center. Thus, various properties of the MOX fuel assembly, such as a void coefficient, an axial power distribution and a reactor shutdown margin, are made substantially equal to those of the uranium fuel assembly.
On the other hand, the following five examples are known in relation to the arrangement of water rods in a uranium fuel assembly:
(3) Fuel Assembly (Japanese Patent Unexamined Publication No. 60-105990)
In this conventional technique, one to three water rods are provided in corner portions of a channel box on the side of a local power range monitor, i.e., on the other side of a control rod, so that errors in the output of the local power range monitor are decreased to thereby improve the load factor and the fuel integrity.
(4) Fuel Assembly and Core of Nuclear Reactor (Japanese Patent Unexamined Publication No. 57-23891)
In this conventional technique, one to five water rods are provided in corner portions of a channel box on the side of a control rod, so that the thermal duty of fuel rods adjacent to the control rod is decreased so as to prevent generation of cracks.
(5) Fuel Assembly (Japanese Patent Unexamined Publication No. 57-583)
In this conventional technique, water rods are provided in corner portions of a channel box on the side of a control rod, to thereby prevent a decrease in the thermal margin owing to the control rod history effect.
(6) Fuel Assembly (Japanese Patent Unexamined Publication No. 60-201284)
In this conventional technique, one part of an upper portion of a fuel rod is formed into a water rod, and such fuel rods are provided in corner portions of a channel box, so as to improve the reactor shutdown margin and flatten the axial power distribution.
(7) Nuclear Reactor (Japanese Patent Unexamined Publication No. 60-222791)
In this conventional technique, the nuclear reactor comprises fuel assemblies in which the number of water rods provided in each fuel assembly is increased from the center of a reactor core toward the periphery, and there are located in the outermost periphery of the reactor core fuel assemblies in which water rods are provided on the diagonal lines including corner portions of a channel box, to thereby uniform the exposure distribution in the nuclear reactor and to save uranium.
The above-described conventional example (1) involves the following problems. That is to say, since the number of water rods is increased and the number of fuel rods is decreased, the average linear heat generation ratio is increased. Therefore, the local power peaking must be further lowered to observe the limit value of the maximum linear heat generation ratio. Consequently, the design of a fuel enrichment distribution becomes more complicated. Further, an increase in the number of water rods and a decrease in the number of MOX fuel rods result in a decrease in the plutonium load per fuel assembly.
In the conventional example (2), the void reactivity coefficient of the mixed type fuel assembly is improved. However, the value is still about -9.5 [%K/K/%void]. When it is compared with the value of the uranium fuel assembly which is about -8.3 [%K/K/%void], it can be understood that the improvement effect is insufficient.
Moreover, when the known techniques in relation to the arrangement of water rods described in the conventional examples (3) to (7) are applied to the means for improving the neutron moderation, the following problems are induced:
In application of the conventional examples (3) to (5), locations of the water rods are asymmetric so that the power peaking will be increased.
In application of the conventional example (6), only the part of the upper portion of each fuel rod is formed into a water rod, and consequently, the neutron moderation is not adequately improved.
In application of the conventional example (7), the fuel assemblies are special ones which are located in the outermost peripheral range of the reactor core and can not be provided for general use. Further, the disclosed technique is concerned with uranium fuel alone, and MOX fuel is not mentioned.
In other words, when this conventional example (7) having more water rods and less fuel rods is compared with a fuel assembly which produces the same power, the fuel rod average heat generation ratio, i.e., the heat generation ratio per unit length of a fuel rod, is larger. In order to avoid this situation, it is suggested to increase the number of water rods in assemblies in the outer peripheral portion of the reactor core which produce low power. Because they are the assemblies which can only be used in the outer peripheral portion of the reactor core, they can not be provided for general use.
Furthermore, when a water rod is provided in the second layer from the outermost periphery and located adjacent to a water rod in each corner portion of the fuel assembly, the local power of fuel rods in the outer peripheral portion which are in contact with these two water rods is increased, and it is not favorable from the standpoint of limitation of the linear heat generation ratio. Therefore, location of such fuel assemblies is limited to the outer peripheral portion of the reactor core.
The above-described seven problems are concerned with the BWR. On the other hand, in a pressurized water reactor (PWR), the core is not boiling during normal operation so that the void reactivity coefficient as in the case of the BWR is not a problem. However, there is a moderator temperature coefficient serving as an index for-indicating a reactivity change with respect to a water density change. When MOX fuel is used, this moderator temperature coefficient is deteriorated, similarly to the void reactivity coefficient in the case of the BWR. In order to improve this moderator temperature coefficient, it is effective to provide water rods without decreasing the number of MOX fuel rods largely, similarly to the improvement of the void reactivity coefficient in the case of the BWR.